The invention relates to the field of corrosion testing of zirconium alloys used in nuclear reactors such as pressurized water reactors (PWR) and pressurized heavy water reactors (PHWR) for components and members which are exposed during reactor operation to neutron flux and the hot aqueous coolant under pressure.
Testing in an autoclave of specimens of a zirconium alloy which correspond to the zirconium alloy of actual reactor members is not new. See, for example, U.S. Pat. No. 4,440,862, issued Apr. 3, 1984. The problem of prior art tests has been the required duration, often months and even years, that are required for meaningful results.
Accordingly, it is an important object of the invention to provide an improved short-term autoclave test for ex-reactor evaluation of in-reactor corrosion resistance of zirconium alloy members for use in pressurized water reactors and pressurized heavy water reactors.
In performing the tests, the American Society for Testing Materials (ASTM) "Standard Test Method for Corrosion Testing of Products of Zirconium, Hafnium, and Their Alloys in Water at 680.degree. F. or in Steam at 750.degree. F.," (Designation:G2-88) is used to the extent it is not specifically varied by the claimed procedure. For instance, to insure the simulated coolant is of "low-oxygen", observance by the new procedure of the 45 part per billion oxygen content limit of paragraph 13.1 of G2-88 and venting of the autoclave in the preliminary procedure prior to the actual test (paragraph 14.3) of G2-88, are used.
In order to more quickly measure the corrosion of the specimen and compare the value with values from tests of other specimens to evaluate the in-reactor corrosion resistance of the zirconium alloy member to which the specimen corresponds relative to the in-reactor corrosion resistance of the zirconium alloy members to which the other specimens correspond, it was necessary to establish what happens to cause corrosion of a zirconium alloy member, such as fuel rod cladding tube, in the reactor during operation. A recognition that in-reactor corrosion resistance of zirconium alloys at extended burnups is degraded due to the fracture of hydride precipitates at the metal-oxide interface, has led to an understanding of the reactions in a nuclear reactor.
The aqueous corrosion of zirconium alloys generates hydrogen as a result of the oxidation reaction. A fraction of this generated hydrogen is absorbed by the alloy (metal). When the hydrogen concentration in the metal exceeds the hydrogen solubility limit associated with the corrosion reaction temperature, hydride precipitation occurs. As a result of the volume expansion associated with the oxidation reaction, the corrosion reaction subjects the metal layer to a tensile stress and the oxide layer to a compressive stress. If the hydride precipitates in the metal are brittle at the corrosion temperature, they are unable to withstand the tensile stress generated by the oxidation process and fracture. The fracture of the hydride precipitates disturb the coherency of the metal-oxide interface and renders the oxide sub-layer non-protective. As a result, hydride precipitation enhances the corrosion rate of zirconium alloys. This type of acceleration of corrosion rate occurs in low-oxygen coolant nuclear reactors and was not recognized by earlier investigations. This discovery of the corrosion rate enhancement for zirconium alloys due to hydride precipitation and subsequent fracture forms the basis of the current invention.
Hydrogen has a tendency to migrate towards cooler parts of the zirconium alloy components. Because of imposed heat flux on a fuel cladding in the reactor, the coolest part of the cladding is adjacent to the barrier oxide layer. As a result, the hydrides are concentrated next to the barrier layer. This is not the case for inreactor components without an imposed heat flux or in isothermal autoclave operation where hydrides are uniformly distributed in the cross-section.
The zirconium hydride is a brittle phase at temperatures lower than 427.degree. C. and above this temperature the zirconium hydride phase exhibits some ductility. Therefore, for metal-oxide interface temperatures less than 427.degree. C., the brittle zirconium hydride phase cannot withstand the tensile strain imposed on the substrate metal by the newly forming zirconium oxide (Zr to ZrO.sub.2 reaction involves a 56% expansion) and the hydride fractures. Such fracture of zirconium hydride destroys the coherency at the metal-oxide interface, thereby, decreases the "protective" nature of the barrier oxide layer which results in an increased corrosion rate. This is probably the reason for the enhanced in-PWR corrosion observed with hydrogen uptake close to the solubility limit, the "thick-film" hypothesis proposed by Johnson in D. D. Lanning, A. B. Johnson, Jr., D. J. Trimble and S. M. Boyd, "Corrosion and Hydriding of N Reactor Pressure Tubes", Zirconium in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, L. F. P. Van Swam and C. M. Eucken Eds., American Society for Testing and Materials, Philadelphia, 1989, pp 3-19.
For metal-oxide interface temperatures greater than 427.degree. C., the hydride phase is ductile and with increasing temperatures it can withstand the strains imposed by the oxide layer more effectively. Therefore, zirconium hydride precipitates are not principal reasons for corrosion rate enhancement at higher temperatures (&gt;427.degree. C.).
The long term (&gt;300 days) rate transition observed in prior art autoclave corrosion tests is also related to the hydride precipitation. However, due to the absence of the heat flux, hydride precipitation does not preferentially occur near the metal-oxide interface. Accordingly, long autoclave times are necessary to charge the entire tube wall cross-section to observe the effect of brittle hydrides on the coherency of the metal-oxide interface.